The present invention relates to a method of vitrifying a high-level radioactive liquid waste generated in the step of reprocessing spent nuclear fuels. More particularly, the present invention is concerned with a vitrification method by which a vitrified waste having a high waste content can be obtained.
A high-level radioactive liquid waste (hereinafter referred to simply as "high-level liquid waste",) is generated in the step of separating U and Pu by reprocessing spent nuclear fuels generated in nuclear power stations. This high-level liquid waste contains various components such as fission products contained in spent nuclear fuels in the form of a solution in nitric acid or a precipitate in a nitric acid medium without being dissolved. Further, the high-level liquid waste contains Na added as a reagent in the reprocessing step and also Fe, Cr and Ni which are corrosion products.
Such a high-level liquid waste is mixed with a raw glass material consisting mainly of SiO.sub.2 and B.sub.2 O.sub.3 in a glass melting furnace at high temperatures and melt-solidified into a vitrified waste. In this process, the nitrate component in the high-level liquid waste is removed in the form of steam and NO.sub.x while the fission products are homogeneously mixed with the raw glass material and vitrified. The resultant vitrified waste is stored for cooling for 30 to 50 years and thereafter disposed of in a stratum more than hundreds of meters deep underground.
Table 1 gives some examples of the chemical compositions of raw glass materials conventionally employed in the vitrification of a high-level liquid waste by Power Reactor and Nuclear Fuel Development Corporation (Doryokuro Kakunenryo Kaihatsu Jigyodan) who is the assignee of the present invention.
TABLE 1 ______________________________________ Examples of chemical compositions of conventional raw glass materials [unit: wt. %] Designation of raw glass material compsn. component PF500 PF606 PF798 ______________________________________ SiO.sub.2 61.83 68.52 62.30 B.sub.2 O.sub.3 20.18 19.60 19.00 Al.sub.2 O.sub.3 5.04 3.50 6.70 CaO 2.88 1.39 4.00 ZnO 2.88 1.39 4.00 Li.sub.2 O 4.32 2.80 4.00 miscellaneous 2.88 2.79 0.00 component ratio B.sub.2 O.sub.3 /SiO.sub.2 0.33 0.29 0.31 ZnO/Li.sub.2 O 0.67 0.50 1 Al.sub.2 O.sub.3 /Li.sub.2 O 1.17 1.25 1.68 ______________________________________
In the conventional vitrification, the waste such as fission products and the raw glass material are mixed generally in proportions of about 25% (on the basis of oxide weight, the same shall apply hereinafter) of the waste and about 75% of the raw glass material. That is, the raw glass material is contained in the vitrified waste in an amount about thrice greater than that of the waste components such as fission products to be primarily vitrified. This is because, when the waste content is increased while lowering the proportion of the raw glass material, the phenomenon called phase separation occurs such that a water-soluble separated phase composed mainly of Mo which is known as "yellow phase", is separated in the vitrified waste, thereby gravely deteriorating the nuclide confinement performance of the vitrified waste. Further, the fission products contained in the waste generate heat in accordance with their decay, so that an increase in the waste content of the vitrified waste raises the temperature of the central part of the vitrified waste to thereby change the properties of the vitrified waste. This is also the reason for the incapability of increasing the waste content of the vitrified waste.
For highly reducing the volume of the vitrified waste, it is desired to develop a method of vitrifying a high-level liquid waste in which, even if the waste content of the vitrified waste is increased over the conventional level of about 25%, the same leaching rate as that of the conventional vitrified waste is ensured without suffering from the yellow phase separation.